National Analytical Management Program (NAMP) U.S. Department of Energy Carlsbad Field Office Radiochemistry Webinars Nuclear Fuel Cycle Series The PUREX Process In Cooperation with our University Partners
2 Meet the Presenter Dr. Jimmy T. Bell Dr. Jimmy T. Bell received his BA degree in mathematics and chemistry from Berry College in Rome, Georgia, and his Ph. D. in physical chemistry from the University of Mississippi. In 2011, Dr. Bell received the Distinguished Alumni Achievement Award from Berry College. After receiving his doctorate, Dr. Bell worked for 31 years at the Oak Ridge National Laboratory, where he retired as Head of the Chemical Development Section of the Chemical Technology Division, in charge of reprocessing and waste treatment activities. He has conducted studies for the U.S. Department of Energy and the U.S. Nuclear Regulatory Commission on nuclear waste separations technologies, fission product release, and nuclear fuel reprocessing technologies, as well as security-sensitive nuclear proliferation reviews for the State Department. For ten years he served as Co-Chairman of the Biannual International Symposium on Separation Science and Technology. He has published more than 45 peer-reviewed papers on subjects relating to nuclear processing, actinide separations technologies, and nuclear waste management. In 2012, Dr. Bell was a recipient of the Glenn T. Seaborg actinide separations award. Phone: 865-376-5408 Email: tennbells@me.com
Used Nuclear Fuel Reprocessing Dr. Jimmy T. Bell Oak Ridge National Laboratory (Retired)
4 Gratitude to: Dr. Raymond G. Wymer Dr. Terry Todd Lucy H. Bell Berta Oates
5 Outline Introduction Used Nuclear Fuel Principal Uses of Uranium and Plutonium Chemistry of Uranium and Plutonium Flowsheet for Processing Used Nuclear Fuel History of the PUREX Process Nuclear Waste Management Containment and Costs
6 Introduction PUREX is a separation technology for removing and recovering U and Pu from used reactor fuels Internationally accepted as best process Solvent extraction process which yields U and Pu products and radioactive waste Separates U and Pu by contacting liquid phase with immiscible phase into which U and Pu transfer U Used fuel HNO 3 Solvent Dissolution Solution U, Pu, FP Extraction Pu FP
7 Spent Nuclear Fuel What is it? Cs and Sr 0.3% Other Long-Lived Fission Products 0.1 % Long-lived I and Tc 0.1% Uranium 95.6% Plutonium 0.9 % Other Minor Actinides 0.1% Stable Fission Products 2.9% Without cladding Most heat production is from Cs and Sr, which decay in ~300 yr Most radiotoxicity is in long-lived fission products and the minor actinides, which can be transmuted and/or disposed in much smaller packages Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
8 Characteristics of Used Fuels Consist primarily of uranium <1% plutonium Contain radioactive fission products Contain stable fission products Very radioactive and require remote handling Very thermally hot
9 Decay of Radioactivity of Fission Products in One Tonne of Spent PWR Fuel Slide courtesy of Dr. R. G. Wymer
10 Uranium and Plutonium Production In the 1940 s these were needed for weapons production Post World War II interest included peaceful uses Nuclear reactors converted to generating steam for electricity production
11 Major Postwar Reactors Primarily used for power production and to power naval vessels Light Water Reactors (BWRs) have one steam cycle Pressurized Water Reactors (PWRs) have two steam cycles Candu reactor is heavy-water cooled These reactors are fueled by uranium
12 Other Reactor Designs Mixed-oxide (MOX) fuels recently used in LWRs Mixed-oxide fuels are 95% U and 5% Pu Breeder reactor concept developed 1960-1980-present Breeder reactors use MOX fuels that are 82% U and 18% Pu Breeder reactors require used fuel processing
13 Fundamental Chemistry of PUREX Used fuel dissolves in nitric acid HNO 3 is compatible with stainless steel U + HNO 3 UO 2 (NO 3 ) 2 + NO x UO 2 + HNO 3 UO 2 (NO 3 ) 2 + H 2 O + NO x Pu + HNO 3 Pu(NO 3 ) 4 + NO x PuO 2 + HNO 3 Pu(NO 3 ) 4 + H 2 O + NO x U and Pu react with phosphate ligands of TBP Form complexes soluble in organic phase
14 PUREX Process Flow Chart Slide courtesy of Dr. R. G. Wymer
15 THORP Process Flowchart Slide courtesy of Dr. R.G. Wymer
16 Picture of Processing Plant Sellafield UK
17 Used Nuclear Reactor Fuel Arrives at processing in shipping casks Placed in circulating water pool called Fuel Receiving Pool (FRP) for storage prior to processing FRP is adjacent to processing canyon and is constructed of reinforced concrete with large overhead cranes to move the fuel canisters
18 Fuel Assembly Source: Syeilendra Pramuditya on April 14, 2009
19 Fuel Disassembly All hardware of the fuel assembly is removed Fuel rods are grouped for shearing
20 Fuel Shearing Most support metals are removed For LWR fuels, rods in a fuel bundle may be sheared For U metal fuels, fuel rods are separated from supports and fed into nitric acid Volatile off-gases released are trapped in filters or scrubbers
21 Fuel Decladding Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
22 Fuel Dissolution and Clarification Dissolvers can be batch operated or continuous Fuel quantities controlled to avoid criticality Neutron poisons added to avoid criticality Batch operations may be 3 tubs or a console Continuous operations may be wheel-like Considerable heat generated; water cooling is necessary
23 Dissolution/Feed Clarification Nitric acid dissolves UO2 pellet from cladding hull, forming UO2(NO3)2 in solution Dissolver product contains approx. 300 g/l uranium Releases radioactive off-gas (iodine, krypton, xenon, carbon- 14, small amounts of tritium) Undissolved solids are removed by centrifugation before transfer to solvent extraction process Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
24 THORP Process Flowchart Slide courtesy of Dr. R.G. Wymer
25 PUREX Process Flow Chart Slide courtesy of Dr. R. G. Wymer
26 PUREX Process- Basic principles Extraction Scrubbing Stripping Organic Solvent b a a M M M b M b a Feed Strip Scrub Solution Separates metal to be recovered Removes impurities from metal Recovers product in solution
27 Basic Solvent Extraction TBP is added TBP Complex Organic Solvent 1) Mix Phases UO 2 +2 Pu 4+ UO 2 +2 UO 2 +2 Cs + Sr 2+ FP UO 2 +2 FP 2) Allow to Settle Cs + Sr 2+ FP FP Pu 4+ FP Am3+ FP Am 3+ Aqueous Solution UO2 2+ + 2NO 3 + 2TBP Pu 4+ + 4NO 3 + 2TBP UO2(NO 3 ) 2 2TBP Pu(NO 3 ) 4 2TBP Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
28 Solvent Extraction Contactors Mixer-Settlers require plenty of floor space Pulse Columns are tall and require head space Centrifugal are smallest units, highest throughput
29 Mixer Settlers Discrete stage units (with efficiencies < 1) Low capital cost Requires large amount of floor space (but low headroom) Large solvent inventory Long residence times Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
Pulse Extraction Column 30 Several feet of column needed for one theoretical stage Low capital cost Requires large amount of head space (40-50 ), but little floor space Moderate solvent inventory Long residence times Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
31 Centrifugal Contactors Each unit near one theoretical stage Higher capital cost Requires little headroom or floor space, but requires remote maintenance capability Small solvent inventory Short residence times Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
32 PUREX Process Basic Principles Tri-butyl phosphate forms soluble complexes with uranyl nitrate and plutonium nitrate (neutral species of U(VI) and Pu(IV)) Spent fuel is dissolved in nitric acid and is then mixed with a solution of TBP in a hydrocarbon diluent (immiscible with aqueous phase) At higher nitric acid concentrations (>0.5 M) the plutonium and uranium partition to the organic (solvent) phase while most of the metals and fission products stay in the aqueous phase Once separated from the fission products, the solvent can be mixed with another aqueous solution of low acidity (<0.01 M) and the uranium and plutonium will partition back to the aqueous phase. To separate plutonium from uranium, a reductant is added to the aqueous stream, reducing Pu(IV) to Pu(III), which is not soluble in the organic solvent and partitions to the aqueous phase while U(VI) remains in the solvent Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
33 PUREX Basic Principles (cont.) TBP is an effective extractant, but is too dense and viscous to use pure Hydrocarbon diluent used to improve physical characteristics Typically 30 vol% TBP is used in the PUREX process Diluents typically dodecane or kerosene (straight or branch chain hydrocarbons ranging from C-10 to C-14) Salting effect Uranium and plutonium extraction is a function of nitrate concentration (called salting effect) Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
34 PUREX Process Advantages and Disadvantages Advantages of liquid-liquid extraction Continuous operation/ High throughput Countercurrent operation/ High purity and selectivity Recycle solvent, minimizing waste Disadvantages of liquid-liquid extraction Solvent degradation due to hydrolysis and radiolysis Degradation products interfere with process chemistry Dibutyl and monobutyl phosphates Efficiently extract Pu, but cannot strip Pu from DBP or MBP Requires substantial tankage and reagents Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
Picture of Processing Plant 35
36 PUREX Process Waste LIQUIDS HLW (RAFFINATE FROM FIRST CYCLE TANK WASTE) LAW (SOLVENT SCRUB; EVAPORATORS) GASES 85 Kr (DISSOLVER OFF-GAS; UNTREATED IN THE PAST) 129 I (DISSOLVER OFF-GAS; REMOVED FROM EARLIEST DAYS) 14 C (AS CO2) (DISSOLVER OFF-GAS; UNTREATED IN THE PAST) 3 H (MOSTLY AS TRITIATED WATER VAPOR) SOLIDS HLW (CONTAMINATED EQUIPMENT; CLADDING HULLS?) LAW (MISCELLANEOUS WASTES FROM OPERATIONS) Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
37 After Primary Extraction U and Pu are in the organic phase U and Pu must then be separated Aqueous phase contains HLW products HLW transferred from PUREX canyon to vitrifier Transfer line must have containment Short HLW transfer line for safety
38 THORP Process Flowchart Slide courtesy of Dr. R.G. Wymer
39 PUREX Process Flow Chart Slide courtesy of Dr. R. G. Wymer
40 Product Conversion Uranyl nitrate is converted to UO3 by denitration at elevated temperature Produces NOx off-gas Plutonium nitrate is precipitated by oxalate or peroxide and calcined to PuO2 Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
41 History of reprocessing in the US J. Schlueter, 2009 Fuel Cycle Information Exchange, June 2009
42 History of PUREX In 1942, U. S. at war with Nazi regime and Japan The Manhattan Project developed to build a uranium weapon, requiring enriched U Natural U contains only 0.7% 235 U Technology for U enrichment had to be developed
43 Building a Plutonium Weapon Required neutron irradiation 238 U + neutron 239 Np 239 Pu Nuclear reactors built to produce used fuel containing <1% 239 Pu Process had to be developed to recover the 239 Pu from the used reactor fuel
44 Uranium Enrichment, 1942-1945 Highly enriched 235 U (>20%) required for weapons Several separation technologies developed Electromagnetic (Oak Ridge) Gaseous diffusion (Oak Ridge Thermal diffusion (Oak Ridge)
45 Reactor Development Fermi pile ---Stag Field, Chicago Helium-cooled pile ---Argonne Oak Ridge air-cooled pile---graphite Reactor Hanford pile included water cooling
46 Evolution of Reprocessing/Recycling 1 st Generation (1940s -1950s) Hanford - USA (Bismuth process) Sellafield - UK (Butex process) 2 nd generation (1960s) Sellafield - UK (Magnox Plant, Purex Process) Marcoule France (UP1, Purex Process) Tokai-Mura Japan (Purex Process) Mayak Russia (Purex Process) 3 rd Generation (1960s 2010) Sellafield UK (Thorp) La Hague France (UP2-800, UP3) Rokkasho Mura -Japan Mixed Oxide Fuel Fabrication Facility (MFFF) - USA 4 th Generation (>2010) Minor Actinides Recycling THORP Uranium Oxide plant Sellafield, UK Sellafield UK UP2 and UP3, La Hague, France
Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008 47 Aqueous Processing - History Began during Manhattan Project to recover Pu-239 Seaborg first separated microgram quantities of Pu in 1942 using bismuth-phosphate precipitation process Process scaled to kilogram quantity production at Hanford in 1944 A scale-up factor of 10 9!!! Solvent extraction processes followed to allow concurrent separation and recovery of both U and Pu and Reprocessing transitioned from defense to commercial use Focus on recycle of uranium and plutonium Waste management Hanford T-Plant 1944 20 micrograms of plutonium hydroxide 1942
48 Bismuth Phosphate Process Dissolution of irradiated fuel or targets in nitric acid Pu valance adjusted to Pu (IV) with sodium nitrite Add sodium phosphate and bismuth nitrate Pu (IV) precipitates as Pu3(PO4)4 PPT re-dissolved in nitric acid, oxidized to Pu (VI), then re-ppt BiPO4 to decontaminate Pu from fission products Recover Pu by reducing to Pu (IV) and re- ppt Repeat cycles w/ LaF to further decontaminate Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
49 Bismuth Process (cont.) Advantages of Bismuth Phosphate Process Recovery of >95% of Pu Decontamination factors from fission products of 107 Disadvantages of Bismuth Phosphate Process Batch operations Inability to recovery uranium Required numerous cycles and chemicals Produced large volumes of high-level waste Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
50 REDOX Process First solvent extraction process used in reprocessing Continuous process Recovers both U and Pu with high yield and high decontamination factors from fission products Developed at Argonne National Laboratory Tested in pilot plant at Oak Ridge Nat. Lab 1948-49 REDOX plant built in Hanford in 1951 Used at Idaho for U-235 recovery Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
51 REDOX Process Hexone (methyl isobutyl ketone) used as the extractant Immiscible with water Used to purify uranium ore concentrates Extracts both uranyl and plutonyl nitrates selectively from fission products Plutonium oxidized to Pu (VI) for highest recovery U (VI) and Pu (VI) co-extracted, then Pu is reduced to Pu (III) by ferrous sulfamate and scrubbed from the solvent Hexone is highly flammable and volatile Large amounts of nonvolatile salt reagents added to process increased waste volume Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
52 PUREX Process Tri-n butyl phosphate used as the extractant in a hydrocarbon diluent (dodecane or kerosene) Suggested by Warf in 1949 for the recovery of Ce (IV) from rare earth nitrates Developed by Knolls Atomic Power Lab. and tested at Oak Ridge in 1950-52 Used for Pu production plant at Savannah River in 1954 (H-canyon facility still operational in 2008) Replaced REDOX process at Hanford in 1956 Modified PUREX used in Idaho beginning in 1953 (first cycle) Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
53 Advantages of PUREX over REDOX Process Nitric acid is used as salting and Advantages scrubbing of PUREX agent and over can be REDOX process evaporated results in less HLW TBP is less volatile and flammable than hexone TBP is more chemically stable in a nitric acid environment Operating costs are lower Nitric acid is used as salting and scrubbing agent and can be evaporated results in less O O HLW O TBP is less volatile and flammable than hexone TBP is more chemically stable in a nitric acid environment Operating costs are lower O P Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
Commercial History in US West Valley, NY First plant in US to reprocess commercial SNF Operated from 1966 until 1972 Capacity of 250-300 MTHM/yr Shutdown due to high retrofit costs associated with changing safety and environmental regulations and construction of larger Barnwell facility Morris, IL Construction halted in 1972, never operated Close-coupled unit operations with fluoride volatility polishing step (dry U feed) Barnwell, SC 1500 MTHM capacity Construction nearly completed- startup testing was in progress 1977 change in US policy on reprocessing stopped construction Plant never operated with spent nuclear fuel West Valley, NY Barnwell, SC 54
55 Commercial Reprocessing in Non-US France Magnox plant in Marcoule began operation in 1958 (~400 MT/yr) Magnox plant in La Hague began operation in 1967 (~400 MT/yr) LWR oxide plant (UP2) began in La Hague in 1976 (800 MT/yr) LWR oxide plant (UP3) began in La Hague in 1990 (800 MT/yr) United Kingdom Windscale plant for Magnox fuel began in 1964 (1200-1500 MT/yr) THORP LWR oxide plant began in 1994 (~1200 MT/yr) Japan Tokai-Mura plant began in 1975 (~200 MT/yr) Rokkasho plant currently undergoing hot commissioning (800 MT/yr) Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
56 PUREX Process Advantages and Disadvantages (History Cont.) Advantages of liquid-liquid extraction Continuous operation/ High throughput Countercurrent operation/ High purity and selectivity Recycle solvent, minimizing waste Disadvantages of liquid-liquid extraction Solvent degradation due to hydrolysis and radiolysis Degradation products interfere with process chemistry Dibutyl and monobutyl phosphates Efficiently extract Pu, but cannot strip Pu from DBP or MBP Requires substantial tankage and reagents Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
57 PUREX Process (History) Used internationally for past 60 years Most advantageous No competitive process apparent for now
58 PUREX-Generated Wastes HLW aqueous raffinate from first-cycle contactors: ~1 Ci/liter requires glass storage Intermediate Level (ILW) not defined by U.S. Nuclear Regulatory Commission Additional processing converts ILW to HLW and LLW Low-level waste (LLW) may be contact-handled Grout storage
59 1960-1980 Fuel Cycle Development HLW management: borosilicate glassification, repositories Used reactor fuel reprocessing stopped in 1978 Yucca Mountain Geological Repository funding discontinued in 2010
60 Used Nuclear Fuel in the US 72 plant sites with spent fuel + DOE sites 39 states with spent fuel 69,500 metric tons of spent fuel currently in US 135,000 metric tons of spent fuel projected by 2035 (excluding new builds)
US Fuel Cycle Policy 61 The current US approach is a oncethrough fuel cycle It is likely that most of the ~70,000 MT of current used fuel will be directly disposed in a geologic repository, beginning around mid-century (~2048) The US is evaluating advanced (closed) fuel cycles for potential deployment also around mid-century No decisions have been made on future fuel cycles in the US, nor which technologies will be employed, if a closed fuel cycle is selected The closed fuel cycle, if implemented, would likely process future generated UNF At current UNF generation rates, 2000 MT/yr, there will be adequate fuel available
Civilian Spent Nuclear Fuel (MTHM Cumulative ) 62 Used Fuel Inventory in US 150,000 Technical Capacity 120,000 90,000 60,000 Current Nuclear Energy Generation Statutory Capacity of First Repository 30,000 2015 0 1980 1990 2000 2010 2020 2030 2040 2050 Year Slide courtesy of Dr. Terry Todd, Nuclear Regulatory Commission Seminar, March 25, 2008
63 Blue Ribbon Committee in 2012 No reconsideration of Yucca Mountain Choose perhaps three sites for air-cooled temporary storage
64 Decay of Radioactivity of Fission Products in One Tonne of Spent PWR Fuel Slide courtesy of Dr. R. G. Wymer
65 PUREX Process Flow Chart Slide courtesy of Dr. R. G. Wymer
Picture of Processing Plant 66
Reducing Costs for U.S. Used Fuel Processing 67 Long-term storage (>60 yrs) reduces fuel activity by two half-lives of Cs-137 and Sr-90 Wall thickness of process canyon could be reduced Fuel receiving area could be air-cooled Low-activity areas need not be located in heavily shielded process canyon Centrifugal contactors could be used
68 Results of Processing Older Used Fuel Cost of building process canyon could then be reduced by ~50% of cost for handling typical 3-years cooled waste
Upcoming Webinars in the Nuclear Fuel Cycle Series Advanced Partitioning Technologies in the U.S. Advanced Partitioning Technologies in Europe Radiation Chemistry at the Back End of the Nuclear Fuel Cycle NAMP website http://www.wipp.energy.gov/namp/